T. Braunroth, N. Berner, F. Rowold, M. P'eridis, M. Stuke
Cosmic-ray muons can be used for the non-destructive imaging of spent nuclear fuel in sealed dry storage casks. The scattering data of the muons after traversing provides information on the thereby penetrated materials. Based on these properties, we investigate and discuss the theoretical feasibility of detecting single missing fuel rods in a sealed cask for the first time. We perform simulations of a vertically standing generic cask model loaded with fuel assemblies from a pressurized water reactor and muon detectors placed above and below the cask. By analysing the scattering angles and applying a significance ratio based on the Kolmogorov-Smirnov test statistic we conclude that missing rods can be reliably identified in a reasonable measuring time period depending on their position in the assembly and cask, and on the angular acceptance criterion of the primary, incoming muons.
{"title":"Muon radiography to visualise individual fuel rods in sealed casks","authors":"T. Braunroth, N. Berner, F. Rowold, M. P'eridis, M. Stuke","doi":"10.1051/EPJN/2021010","DOIUrl":"https://doi.org/10.1051/EPJN/2021010","url":null,"abstract":"Cosmic-ray muons can be used for the non-destructive imaging of spent nuclear fuel in sealed dry storage casks. The scattering data of the muons after traversing provides information on the thereby penetrated materials. Based on these properties, we investigate and discuss the theoretical feasibility of detecting single missing fuel rods in a sealed cask for the first time. We perform simulations of a vertically standing generic cask model loaded with fuel assemblies from a pressurized water reactor and muon detectors placed above and below the cask. By analysing the scattering angles and applying a significance ratio based on the Kolmogorov-Smirnov test statistic we conclude that missing rods can be reliably identified in a reasonable measuring time period depending on their position in the assembly and cask, and on the angular acceptance criterion of the primary, incoming muons.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":null,"pages":null},"PeriodicalIF":0.5,"publicationDate":"2021-02-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47829801","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Prediction of crack propagation kinetics in the components of nuclear plant primary circuits undergoing Stress Corrosion Cracking (SCC) can be improved by a refinement of the SCC models. One of the steps in the estimation of the time to rupture is the crack propagation criterion. Current models make use of macroscopic measures (e.g. stress, strain) obtained for instance using the Finite Element Method. To go down to the microscopic scale and use local measures, a two-step approach is proposed. First, synthetic microstructures representing the material under specific loadings are simulated, and their quality is validated using statistical measures. Second, the shortest path to rupture in terms of propagation time is computed, and the distribution of those synthetic times to rupture is compared with the time to rupture estimated only from macroscopic values. The first step is realized with the cross-correlation-based simulation (CCSIM), a multipoint simulation algorithm that produces synthetic stochastic fields from a training field. The Earth Mover’s Distance is the metric which allows to assess the quality of the realizations. The computation of shortest paths is realized using Dijkstra’s algorithm. This approach allows to obtain a refinement in the prediction of the kinetics of crack propagation compared to the macroscopic approach. An influence of the loading conditions on the distribution of the computed synthetic times to rupture was observed, which could be reduced through a more robust use of the CCSIM.
{"title":"Prediction of crack propagation kinetics through multipoint stochastic simulations of microscopic fields","authors":"Etienne Le Mire, E. Burger, B. Iooss, C. Mai","doi":"10.1051/EPJN/2021001","DOIUrl":"https://doi.org/10.1051/EPJN/2021001","url":null,"abstract":"Prediction of crack propagation kinetics in the components of nuclear plant primary circuits undergoing Stress Corrosion Cracking (SCC) can be improved by a refinement of the SCC models. One of the steps in the estimation of the time to rupture is the crack propagation criterion. Current models make use of macroscopic measures (e.g. stress, strain) obtained for instance using the Finite Element Method. To go down to the microscopic scale and use local measures, a two-step approach is proposed. First, synthetic microstructures representing the material under specific loadings are simulated, and their quality is validated using statistical measures. Second, the shortest path to rupture in terms of propagation time is computed, and the distribution of those synthetic times to rupture is compared with the time to rupture estimated only from macroscopic values. The first step is realized with the cross-correlation-based simulation (CCSIM), a multipoint simulation algorithm that produces synthetic stochastic fields from a training field. The Earth Mover’s Distance is the metric which allows to assess the quality of the realizations. The computation of shortest paths is realized using Dijkstra’s algorithm. This approach allows to obtain a refinement in the prediction of the kinetics of crack propagation compared to the macroscopic approach. An influence of the loading conditions on the distribution of the computed synthetic times to rupture was observed, which could be reduced through a more robust use of the CCSIM.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":null,"pages":null},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826669","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Since the 80’s, industrial core calculations employ the two-step scheme based on prior cross sections preparation with few energy groups and in homogenized reference geometries. Spatial homogenization in the fuel assembly quarters is the most frequent calculation option nowadays, relying on efficient nodal solvers using a coarse mesh. Pin-wise reaction rates are then reconstructed by dehomogenization techniques. The future trend of core calculations is moving however toward pin-by-pin explicit representations, where few-group cross sections are homogenized in the single pins at many physical conditions and many nuclides are selected for the simplified depletion chains. The resulting data model requires a considerable memory occupation on disk-files and the time needed to evaluate all data exceeds the limits for practical feasibility of multi-physics reactor calculations. In this work, we study the compression of pin-by-pin homogenized cross sections by the Hotelling transform in typical PWR fuel assemblies. The reconstruction of these quantities at different physical states of the assembly is then addressed by interpolation of only a few compressed coefficients, instead of interpolating separately each homogenized cross section. Savings in memory higher than 90% are observed, which result in important gains in runtime when interpolating the few-group data.
{"title":"A multivariate representation of compressed pin-by-pin cross sections","authors":"D. Tomatis","doi":"10.1051/EPJN/2021006","DOIUrl":"https://doi.org/10.1051/EPJN/2021006","url":null,"abstract":"Since the 80’s, industrial core calculations employ the two-step scheme based on prior cross sections preparation with few energy groups and in homogenized reference geometries. Spatial homogenization in the fuel assembly quarters is the most frequent calculation option nowadays, relying on efficient nodal solvers using a coarse mesh. Pin-wise reaction rates are then reconstructed by dehomogenization techniques. The future trend of core calculations is moving however toward pin-by-pin explicit representations, where few-group cross sections are homogenized in the single pins at many physical conditions and many nuclides are selected for the simplified depletion chains. The resulting data model requires a considerable memory occupation on disk-files and the time needed to evaluate all data exceeds the limits for practical feasibility of multi-physics reactor calculations. In this work, we study the compression of pin-by-pin homogenized cross sections by the Hotelling transform in typical PWR fuel assemblies. The reconstruction of these quantities at different physical states of the assembly is then addressed by interpolation of only a few compressed coefficients, instead of interpolating separately each homogenized cross section. Savings in memory higher than 90% are observed, which result in important gains in runtime when interpolating the few-group data.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":null,"pages":null},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826718","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
N. Marie, Simon Li, A. Marrel, M. Marquès, S. Bajard, A. Tosello, Jorge Perez, Baptiste Grosjean, A. Gerschenfeld, M. Anderhuber, Chotaire Geffray, Y. Gorsse, G. Mauger, L. Matteo
Within the framework of the French 4th-generation Sodium-cooled Fast Reactor safety assessment, methodology on VVUQ (Verification, Validation, Uncertainty Quantification) is conducted to demonstrate that the CEA's thermal-hydraulic Scientific Computation Tools (SCTs) are effective and operational for design and safety studies purposes on this type of reactor. This VVUQ-based qualification is a regulatory requirement from the French Nuclear Safety Authority (NSA). In this paper, the current practice of VVUQ approach application for a SFR accidental transient is described with regard to the NSA requirements. It constitutes the first practical, progressively improvable approach. As the SCT is qualified for a given version on a given scenario, the transient related to a total unprotected station blackout has been selected. As it is a very complex multi-scale transient, the SCT MATHYS (which is a coupling of the CATHARE2 tool at system scale, TrioMC tool at component scale and TrioCFD tool at local scale) is used. This paper presents the preliminary VVUQ application to the qualification of this tool on this selected transient. In addition, this work underlines some feedback on design and R&D aspects that should be addressed in the future to improve the SCT.
{"title":"VVUQ of a thermal-hydraulic multi-scale tool on unprotected loss of flow accident in SFR reactor","authors":"N. Marie, Simon Li, A. Marrel, M. Marquès, S. Bajard, A. Tosello, Jorge Perez, Baptiste Grosjean, A. Gerschenfeld, M. Anderhuber, Chotaire Geffray, Y. Gorsse, G. Mauger, L. Matteo","doi":"10.1051/EPJN/2021002","DOIUrl":"https://doi.org/10.1051/EPJN/2021002","url":null,"abstract":"Within the framework of the French 4th-generation Sodium-cooled Fast Reactor safety assessment, methodology on VVUQ (Verification, Validation, Uncertainty Quantification) is conducted to demonstrate that the CEA's thermal-hydraulic Scientific Computation Tools (SCTs) are effective and operational for design and safety studies purposes on this type of reactor. This VVUQ-based qualification is a regulatory requirement from the French Nuclear Safety Authority (NSA). In this paper, the current practice of VVUQ approach application for a SFR accidental transient is described with regard to the NSA requirements. It constitutes the first practical, progressively improvable approach. As the SCT is qualified for a given version on a given scenario, the transient related to a total unprotected station blackout has been selected. As it is a very complex multi-scale transient, the SCT MATHYS (which is a coupling of the CATHARE2 tool at system scale, TrioMC tool at component scale and TrioCFD tool at local scale) is used. This paper presents the preliminary VVUQ application to the qualification of this tool on this selected transient. In addition, this work underlines some feedback on design and R&D aspects that should be addressed in the future to improve the SCT.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":null,"pages":null},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826682","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Léa Tillard, X. Doligez, G. Senentz, M. Ernoult, Jiali Liang, N. Thiollière
This article presents an assessment of fuel cycle parameter impact on waste production through the prism of vitrified container and minor actinide masses, using a scenario simulated with the CLASS code. The number of canister introduces a new focus on waste production estimation for a nuclear fleet, as it helps to set the repository size for deep geological disposal of high level waste. To evaluate the number of canisters, dedicated developments to model a simplified waste vitrification unit in the CLASS package are presented. It relies on artificial neural network estimations of decay heat, α radiation and mass content, for different material flow coming from reprocessing and sent to vitrification. Then, the studied scenario considers a transition from a PWRs plutonium mono-recycling fleet to a plutonium multi-recycling fleet. Vitrified waste container production is calculated as a function of different material reprocessing options. Simulations shows that up to 19% variation may be observed (in 2060) in canisters’ total number depending on the different assumptions. Impact of vitrification parameters such as the size of buffer before vitrification is also analysed and the importance of mixing material coming from MOX and MIX spent fuels with material from UOX spent fuels is clearly established.
{"title":"Estimation of the vitrified canister production for a PWR fleet with the CLASS code","authors":"Léa Tillard, X. Doligez, G. Senentz, M. Ernoult, Jiali Liang, N. Thiollière","doi":"10.1051/epjn/2021020","DOIUrl":"https://doi.org/10.1051/epjn/2021020","url":null,"abstract":"This article presents an assessment of fuel cycle parameter impact on waste production through the prism of vitrified container and minor actinide masses, using a scenario simulated with the CLASS code. The number of canister introduces a new focus on waste production estimation for a nuclear fleet, as it helps to set the repository size for deep geological disposal of high level waste. To evaluate the number of canisters, dedicated developments to model a simplified waste vitrification unit in the CLASS package are presented. It relies on artificial neural network estimations of decay heat, α radiation and mass content, for different material flow coming from reprocessing and sent to vitrification. Then, the studied scenario considers a transition from a PWRs plutonium mono-recycling fleet to a plutonium multi-recycling fleet. Vitrified waste container production is calculated as a function of different material reprocessing options. Simulations shows that up to 19% variation may be observed (in 2060) in canisters’ total number depending on the different assumptions. Impact of vitrification parameters such as the size of buffer before vitrification is also analysed and the importance of mixing material coming from MOX and MIX spent fuels with material from UOX spent fuels is clearly established.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":null,"pages":null},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826609","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Neutron calculations in the neutron shielding of fast neutron reactors are a complex problem as deterministic schemes are usually not suited for such calculations while Monte-Carlo codes have poor computational performance due to the very low flux levels in neutron shields. In this article, both methods are studied, as well as a hybrid scheme on the neutron shielding of the ASTRID fast reactor benchmark. This hybrid scheme uses a fission source calculated by a deterministic code in order to precisely calculate neutron fluxes in the shielding with a Monte-Carlo code using variance reduction techniques. This provides reference results in order to validate deterministic calculations. Comparisons between deterministic codes and this hybrid reference show that large biases are obtained, up to 50%. Further studies are made to reduce the biases, showing that many physical phenomena should be treated, including anisotropy of the scattering law at high energies and spatial self-shielding inside the boron carbide shielding. These improvements reduce the biases to less than 10%. Finally, some applications to designing criteria for the neutron shielding are presented, such as gas production in the neutron shielding and activation of secondary sodium at the intermediate heat exchanger (IHX).
{"title":"A hybrid approach for neutronics calculations in the neutron shielding of sodium fast reactors","authors":"Amine Hajji, C. Coquelet-Pascal, P. Blaise","doi":"10.1051/epjn/2021016","DOIUrl":"https://doi.org/10.1051/epjn/2021016","url":null,"abstract":"Neutron calculations in the neutron shielding of fast neutron reactors are a complex problem as deterministic schemes are usually not suited for such calculations while Monte-Carlo codes have poor computational performance due to the very low flux levels in neutron shields. In this article, both methods are studied, as well as a hybrid scheme on the neutron shielding of the ASTRID fast reactor benchmark. This hybrid scheme uses a fission source calculated by a deterministic code in order to precisely calculate neutron fluxes in the shielding with a Monte-Carlo code using variance reduction techniques. This provides reference results in order to validate deterministic calculations. Comparisons between deterministic codes and this hybrid reference show that large biases are obtained, up to 50%. Further studies are made to reduce the biases, showing that many physical phenomena should be treated, including anisotropy of the scattering law at high energies and spatial self-shielding inside the boron carbide shielding. These improvements reduce the biases to less than 10%. Finally, some applications to designing criteria for the neutron shielding are presented, such as gas production in the neutron shielding and activation of secondary sodium at the intermediate heat exchanger (IHX).","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":null,"pages":null},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826919","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Jiali Liang, M. Ernoult, X. Doligez, S. David, N. Thiollière
During the recent ten years, the estimation of future uranium demands has changed greatly, and SFR competitiveness is called again into question. In this context, a planning of plutonium multi-recycling in PWRs for the near-term decades has been announced in France, which replaces the objective of future SFR deployment. However, the mid-term policy concerning the future reactor system is always uncertain, and the demand of SFR deployment may re-increase significantly. This study looks into this possibility and analyzes the consequences of such back and forth between different plutonium multi-recycling strategies. The newly developed methodology of robustness assessment is applied to the problem, considering the objective disruptions to take into account the deep uncertainties about nuclear future. Two prior trajectories of plutonium multi-recycling, one involving the use of MIX fuel in PWRs and the other considering the SFR deployment, are analyzed first. The disruption of the strategy using MIX is then supposed under the re-consideration of future SFR deployment. To quantify the impacts of using MIX on deployment timing, we investigate the earliest time for which the fleet substitution with SFRs can be completed. To supplement, the prior strategy of SFR deployment is also disrupted under the context of halting the start of new SFR. The plutonium multi-recycling in PWRs, regarded as adaptive strategy, aims to minimize the idle plutonium. In these robustness assessments, numerous outputs of interests are analyzed, in order to provide a comprehensive evaluation of consequences of prior strategies, regarding the uncertain disruptions and optimal readjustments.
{"title":"Impact of disruption between options of plutonium multi-recycling in PWRs and in SFRs","authors":"Jiali Liang, M. Ernoult, X. Doligez, S. David, N. Thiollière","doi":"10.1051/epjn/2021018","DOIUrl":"https://doi.org/10.1051/epjn/2021018","url":null,"abstract":"During the recent ten years, the estimation of future uranium demands has changed greatly, and SFR competitiveness is called again into question. In this context, a planning of plutonium multi-recycling in PWRs for the near-term decades has been announced in France, which replaces the objective of future SFR deployment. However, the mid-term policy concerning the future reactor system is always uncertain, and the demand of SFR deployment may re-increase significantly. This study looks into this possibility and analyzes the consequences of such back and forth between different plutonium multi-recycling strategies. The newly developed methodology of robustness assessment is applied to the problem, considering the objective disruptions to take into account the deep uncertainties about nuclear future. Two prior trajectories of plutonium multi-recycling, one involving the use of MIX fuel in PWRs and the other considering the SFR deployment, are analyzed first. The disruption of the strategy using MIX is then supposed under the re-consideration of future SFR deployment. To quantify the impacts of using MIX on deployment timing, we investigate the earliest time for which the fleet substitution with SFRs can be completed. To supplement, the prior strategy of SFR deployment is also disrupted under the context of halting the start of new SFR. The plutonium multi-recycling in PWRs, regarded as adaptive strategy, aims to minimize the idle plutonium. In these robustness assessments, numerous outputs of interests are analyzed, in order to provide a comprehensive evaluation of consequences of prior strategies, regarding the uncertain disruptions and optimal readjustments.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":null,"pages":null},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826509","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
T. Okamura, R. Katano, A. Oizumi, K. Nishihara, M. Nakase, H. Asano, K. Takeshita
Nuclear Material Balance code version 4.0 (NMB4.0) has been developed through collaborative R&D between TokyoTech&JAEA. Conventional nuclear fuel cycle simulation codes mainly analyze actinides and are specialized for front-end mass balance analysis. However, quantitative back-end simulation has recently become necessary for considering R&D strategies and sustainable nuclear energy utilization. Therefore, NMB4.0 was developed to realize the integrated nuclear fuel cycle simulation from front- to back-end. There are three technical features in NMB4.0: 179 nuclides are tracked, more than any other code, throughout the nuclear fuel cycle; the Okamura explicit method is implemented, which contributes to reducing the numerical cost while maintaining the accuracy of depletion calculations on nuclides with a shorter half-life; and flexibility of back-end simulation is achieved. The main objective of this paper is to show the newly developed functions, made for integrated back-end simulation, and verify NMB4.0 through a benchmark study to show the computational performance.
{"title":"NMB4.0: development of integrated nuclear fuel cycle simulator from the front to back-end","authors":"T. Okamura, R. Katano, A. Oizumi, K. Nishihara, M. Nakase, H. Asano, K. Takeshita","doi":"10.1051/epjn/2021019","DOIUrl":"https://doi.org/10.1051/epjn/2021019","url":null,"abstract":"Nuclear Material Balance code version 4.0 (NMB4.0) has been developed through collaborative R&D between TokyoTech&JAEA. Conventional nuclear fuel cycle simulation codes mainly analyze actinides and are specialized for front-end mass balance analysis. However, quantitative back-end simulation has recently become necessary for considering R&D strategies and sustainable nuclear energy utilization. Therefore, NMB4.0 was developed to realize the integrated nuclear fuel cycle simulation from front- to back-end. There are three technical features in NMB4.0: 179 nuclides are tracked, more than any other code, throughout the nuclear fuel cycle; the Okamura explicit method is implemented, which contributes to reducing the numerical cost while maintaining the accuracy of depletion calculations on nuclides with a shorter half-life; and flexibility of back-end simulation is achieved. The main objective of this paper is to show the newly developed functions, made for integrated back-end simulation, and verify NMB4.0 through a benchmark study to show the computational performance.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":null,"pages":null},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826560","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The generalized perturbation method is described relevant to ratios of bi-linear functionals of the real and adjoint neutron fluxes of critical multiplying systems. Simple linear analysis for optimization and sensitivity studies are then feasible relative to spectrum and space-dependent quantities, such as Doppler and coolant void reactivity effects in fast reactors.
{"title":"The heuristically-based generalized perturbation theory","authors":"A. Gandini","doi":"10.1051/EPJN/2021003","DOIUrl":"https://doi.org/10.1051/EPJN/2021003","url":null,"abstract":"The generalized perturbation method is described relevant to ratios of bi-linear functionals of the real and adjoint neutron fluxes of critical multiplying systems. Simple linear analysis for optimization and sensitivity studies are then feasible relative to spectrum and space-dependent quantities, such as Doppler and coolant void reactivity effects in fast reactors.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":null,"pages":null},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826691","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Among the many domains of reactor physics on which Massimo Salvatores gave his considerable contributions, he was particularly passionate about integral experiments. In this paper, we make a review of selected experimental campaigns among the numerous ones he has promoted, conceived, designed, directed, or analyzed. They have been regrouped in a temporal sequence corresponding to the different periods of Massimo's career, which exceeded 50 years. When possible, for each of the experiments we provide a brief description, the goal for which it was conceived and carried out, and the practical impact on validation and design improvement. Finally, the conclusions offer thoughts and suggestions for the future of the integral experiments and a possible way of honoring the invaluable legacy that Massimo Salvatores has left to us.
{"title":"Massimo Salvatores: integral experiments and their use for the validation of nuclear data and the neutronic design of advanced nuclear systems","authors":"G. Palmiotti, P. Blaise, F. Mellier","doi":"10.1051/epjn/2021007","DOIUrl":"https://doi.org/10.1051/epjn/2021007","url":null,"abstract":"Among the many domains of reactor physics on which Massimo Salvatores gave his considerable contributions, he was particularly passionate about integral experiments. In this paper, we make a review of selected experimental campaigns among the numerous ones he has promoted, conceived, designed, directed, or analyzed. They have been regrouped in a temporal sequence corresponding to the different periods of Massimo's career, which exceeded 50 years. When possible, for each of the experiments we provide a brief description, the goal for which it was conceived and carried out, and the practical impact on validation and design improvement. Finally, the conclusions offer thoughts and suggestions for the future of the integral experiments and a possible way of honoring the invaluable legacy that Massimo Salvatores has left to us.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":null,"pages":null},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826761","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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